IR 05000424/2008003

IR 05000424-08-003, 05000425-08-003, on 04/01/2008 - 06/30/2008, Vogtle Electric Generating Plant, Units 1 and 2, In-service Inspection

ML082130598
Person / Time
Site:	Vogtle
Issue date:	07/31/2008
From:	Scott Shaeffer NRC/RGN-II/DRP/RPB2
To:	Tynan T Southern Nuclear Operating Co
References
IR-08-003
Download: ML082130598 (35)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION II

SAM NUNN ATLANTA FEDERAL CENTER 61 FORSYTH STREET, SW, SUITE 23T85 ATLANTA, GEORGIA 30303-8931 July 31, 2008 Mr. Tom E. Tynan Vice President - Vogtle Southern Nuclear Operating Company, Inc.

Vogtle Electric Generating Plant 7821 River Road Waynesboro, GA 30830 SUBJECT: VOGTLE ELECTRIC GENERATING PLANT - NRC INTEGRATED INSPECTION REPORT 05000424/2008003 AND 05000425/2008003 AND RELEASE OF OFFICE OF INVESTIGATION

SUMMARY

Dear Mr. Tynan:

On June 30, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Vogtle Electric Generating Plant, Units 1 and 2. The enclosed integrated inspection report documents the inspection findings, which were discussed on July 21, 2008, with Mr. C. R.

Dedrickson and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, one finding of very low safety significance was identified which was determined to be a violation of regulatory requirements. Also, a licensee-identified violation, which was determined to be of very low safety significance, is listed in this report. The NRC is treating these violations as a non-cited violation (NCV) consistent with Section VI.A.1 of the NRC Enforcement Policy because of the very low safety significance of the violations and because they are entered into your corrective action program (CAP). If you contest these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk,

Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director,

Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Vogtle Electric Generating Plant.

Also enclosed is the summary of the Nuclear Regulatory Commission (NRC) Office of Investigations (OI) completed report regarding alleged discrimination. OI determined there was insufficient evidence to substantiate the allegation and we plan no further action with regard to this matter.

SNC 2 In accordance with the Code of Federal Regulations 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter, its enclosures, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Scott M. Shaeffer, Chief Reactor Projects Branch 2 Division of Reactor Projects Docket Nos.: 50-424, 50-425 License Nos.: NPF-68 and NPF-81 Enclosures: 1. Inspection Report 05000424/2008003 and 05000425/2008003 w/Attachment: Supplemental Information 2. Factual Summary For OI Report cc w/encl: (See page 3)

_________________________ X SUNSI REVIEW COMPLETE /SMS/

OFFICE RII:DRP RII:DRP RII:DRP RII:DRP RII:DRS RII:DRS RII:DRS SIGNATURE TXL /RA/ GJM /via tele/ GJM /via email PKN /via tele/ ECM /via tele/ RLM /RA/ ADN /RA/

for/

NAME TLighty GMcCoy TChandler PNiebaum EMichel RMoore ANielsen DATE 07/31/2008 07/31/2008 07/31/2008 07/31/2008 07/31/2008 07/31/2008 07/31/2008 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICE RII:DRS SIGNATURE JER /RA/

NAME JRivera DATE 07/31/2008 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO

SNC 3 cc w/encl: Director N. J. Stringfellow Environmental Protection Licensing Manager Department of Natural Resources Southern Nuclear Operating Company, Inc. Electronic Mail Distribution Electronic Mail Distribution Cynthia Sanders Jeffrey T. Gasser Program Manager Executive Vice President Radioactive Materials Program Southern Nuclear Operating Company, Inc. Department of Natural Resources Electronic Mail Distribution Electronic Mail Distribution L. Mike Stinson Jim Sommerville Vice President (Acting) Chief Fleet Operations Support Environmental Protection Division Southern Nuclear Operating Company, Inc. Department of Natural Resources Electronic Mail Distribution Electronic Mail Distribution Michael A. MacFarlane Mr. Steven M. Jackson Southern Nuclear Operating Company, Inc. Senior Engineer - Power Supply 40 Inverness Center Parkway Municipal Electric Authority of Georgia P.O. Box 1295 Electronic Mail Distribution Birmingham, AL 35201-1295 Mr. Reece McAlister David H. Jones Executive Secretary Vice President Georgia Public Service Commission Engineering Electronic Mail Distribution Southern Nuclear Operating Company, Inc.

Electronic Mail Distribution Office of the Attorney General Electronic Mail Distribution Moanica Caston Vice President and General Counsel M. Stanford Blanton, Esq.

Southern Nuclear Operating Company, Inc. Balch and Bingham Law Firm Electronic Mail Distribution Electronic Mail Distribution Resident Manager Office of the County Commissioner Oglethorpe Power Corporation Burke County Commission Alvin W. Vogtle Nuclear Plant Electronic Mail Distribution 7821 River Road Waynesboro, GA 30830 Arthur H. Domby, Esq.

Troutman Sanders Laurence Bergen Electronic Mail Distribution Oglethorpe Power Corporation Electronic Mail Distribution Director Consumers' Utility Counsel Division Mr. N. Holcomb Govenor's Office of Consumer Affairs Commissioner 2 M. L. King, Jr. Drive Department of Natural Resources Plaza Level East; Suite 356 Electronic Mail Distribution Atlanta, GA 30334-4600 Dr. Carol Couch

SNC 4 Susan E. Jenkins Assistant Director, Division of Waste Management Bureau of Land and Waste Management Department of Health and Environmental Control Electronic Mail Distribution

SNC 5 Letter to Tom E. Tynan from Scott M. Shaeffer dated July 30, 2008 SUBJECT: VOGTLE ELECTRIC GENERATING PLANT- NRC INTEGRATED INSPECTION REPORT 05000424/2008003 AND 05000425/2008003 AND RELEASE OF OFFICE OF INVESTIGATION SUMMARY Distribution w/encl:

C. Evans, RII L. Slack, RII OE Mail RIDSNRRDIRS PUBLIC R. Jervey, NRR

U. S. NUCLEAR REGULATORY COMMISSION REGION II Docket Nos.: 50-424, 50-425 License Nos.: NPF-68, NPF-81 Report Nos.: 05000424/2008003 and 05000425/2008003 Licensee: Southern Nuclear Operating Company, Inc. (SNC)

Facility: Vogtle Electric Generating Plant, Units 1 and 2 Location: Waynesboro, GA 30830 Dates: April 1, 2008 through June 30, 2008 Inspectors: G. McCoy, Senior Resident Inspector T. Chandler, Resident Inspector P. Niebaum, Resident Inspector, Hatch E. Michel, Reactor Inspector (Sections 1R08 and 4OA5.4)

R. Moore, Reactor Inspector (Section 4OA5.3)

A. Nielsen, Health Physics Inspector (Sections 2OS1)

J. Rivera, Reactor Inspector (Sections 1R08 and 4OA5.4)

Approved by: Scott M. Shaeffer, Chief Reactor Projects Branch 2 Division of Reactor Projects

SUMMARY OF FINDINGS IR 05000424/2008-003, 05000425/2008-003; 04/01/2008 - 06/30/2008; Vogtle Electric Generating Plant, Units 1 and 2; In-service Inspection The report covered a three-month period of inspection by resident inspectors, reactor inspectors and a health physicist. One Green NCV was identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Rev 4 dated December 2006.

NRC-Identified and Self-Revealing Findings

Green.

A NRC-identified non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, was identified for lack of procedures to address compensatory measures for containment conditions which would mask indications of boric acid leakage. A combination of condensation and leaking containment cooler coils produced a white, crystalline film in large portions of the lower levels of containment that would reduce the adequacy of visuals inspections performed to detect boric acid leakage on susceptible components. The licensee entered the deficiency into their corrective action (CA)program for resolution.

This finding is more than minor because it affected the procedure quality attribute of the Barrier Integrity Cornerstone in that there were no additional measures taken to discriminate between boric acid leakage and the white residue present in containment which masked actual boric acid leakage. The finding is of very low safety significance because no degradation discovered called into question the operability of an affected component. (Section 1R08.3).

Licensee-Identified Violations

Violations of very low safety significance, which were identified by the licensee, have been reviewed by the NRC. Corrective actions taken or planned by the licensee have been entered into the licensee=s CAP. These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 1 started the report period shutdown for a planned refueling outage. The unit was restarted on April 22 and attained full power on April 29. The unit operated at full rated thermal power (RTP) for the remainder of the inspection period.

Unit 2 operated at essentially full RTP for the entire inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

1R01 Adverse Weather Protection

a. Inspection Scope

Grid Reliability. The inspectors reviewed the licensees procedures to verify communication protocols exist between the transmission system operator and the control room to promptly identify issues that could impact the offsite power system. The inspectors verified the adequacy of these procedures to address measures to monitor and maintain availability and reliability of both the offsite alternating current (AC) power system and the alternate AC power system. The inspectors also reviewed the compensatory actions identified in station procedures to be performed when it is not possible to predict post-trip voltage at the site for current electrical grid conditions.

Documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

a. Inspection Scope

Partial System Walkdown. The inspectors performed partial walkdowns of the following three systems to verify correct system alignment. The inspectors checked for correct valve and electrical power alignments by comparing positions of valves, switches, and breakers to the documents listed in the Attachment. Additionally, the inspectors reviewed the condition report database to verify that equipment alignment problems were being identified and appropriately resolved.

C Unit 1 auxiliary feedwater system after the unit was started after a refueling outage C Unit 2 B train diesel generator during an A train diesel generator extended outage C Unit 2 A train auxiliary component cooling water system (ACCW)

Complete System Walkdown. The inspectors performed a complete walkdown of the Unit 2 component cooling water system. The inspectors performed a detailed check of valve positions, electrical breaker positions, and operating switch positions to evaluate the operability of the redundant trains or components by comparing the required position

in the system operating procedure to the actual position. The inspectors also reviewed control room logs, condition reports, and system health reports to verify that alignment and equipment discrepancies were being identified and appropriately resolved. The documents reviewed are listed in the Attachment.

b. Findings

No findings of significance were identified.

1R05 Fire Protection

a. Inspection Scope

Fire Area Tours. The inspectors walked down the following five plant areas to verify the licensee was controlling combustible materials and ignition sources as required by procedures 92015-C, Use, Control, and Storage of Flammable/Combustible Materials, and 92020-C, Control of Ignition Sources. The inspectors assessed the observable condition of fire detection, suppression, and protection systems and reviewed the licensees fire protection Limiting Condition for Operation log and condition report (CR)database to verify that the corrective actions for degraded equipment were identified and appropriately prioritized. The inspectors also reviewed the licensees fire protection program to verify the requirements of Updated Final Safety Analysis Report Section 9.5.1, Fire Protection Program, and Appendix 9A, Fire Hazards Analysis, were met.

Documents reviewed are listed in the Attachment.

C Unit 1 component cooling water (CCW) pump rooms C Unit 1 ACCW pump rooms C Unit 1 rod control switchgear room C Unit 2 rod control switchgear room C Unit 2 auxiliary feedwater pump house

b. Findings

No findings of significance were identified.

1R08 In-service Inspection (ISI) Activities

.1 In-Service Inspection Activities Of The Reactor Coolant System (RCS) Boundary And

Risk Significant Piping Boundaries

a. Inspection Scope

The inspectors reviewed the implementation of the licensees ISI program for monitoring degradation of the reactor coolant system (RCS) boundary and risk significant piping boundaries during the Unit 1 Spring 2008 refueling outage. The inspectors activities consisted of an on-site review of nondestructive examination (NDE) and welding activities to evaluate compliance with the applicable edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Sections II, V, IX, and XI (Code of record for the third 10-year ISI interval was 2001 Edition with 20 Addenda), and to verify that indications and defects (if present) were appropriately evaluated and dispositioned in accordance with the requirements of the ASME Code,Section XI acceptance standards.

The inspectors review of NDE activities consisted of examination procedures, NDE reports, equipment and consumables certification records, personnel qualification records, calibration reports, and calibration block fabrication drawings (as applicable) for the following examinations:

• Ultrasonic (UT) examination of the pressurizer safety nozzle N2 weld overlay as part of Alloy 600 mitigation, Reactor Coolant System (ASME Class 1)

• UT examination of weld 11201-00508, 31 elbow to nozzle weld, Reactor Coolant System (ASME Class 1)

The inspectors review of welding activities included a sample of in process welding activities for ASME Class 1 piping to evaluate compliance with procedures and the ASME Code. The inspectors directly observed part of the welding process and verified welding machine settings for the welding activities described below. The inspector also reviewed weld process control reports, welding procedures, procedure qualification records, certified material test reports for filler material, and welder qualification records.

• Weld Overlay on pressurizer surge nozzle (N6) as part of Alloy 600 mitigation, Reactor Coolant System (ASME Class 1)

b. Findings

No findings of significance were identified.

.2 PWR Vessel Upper Head Penetration (VUHP) Inspection Activities

a. Inspection Scope

The licensee conducted a bare-metal visual examination of vessel upper head in accordance with NRC Order EA-03-009 and its Revision 1. The inspectors reviewed and independently evaluated tapes and photos of the visual examination of numerous CRDM penetrations through the vessel head for evidence of boric acid leakage. The inspectors also reviewed the licensees calculations for effective degradation years (EDYs), as referenced in the Order.

b. Findings

No findings of significance were identified

.3 Boric Acid Corrosion Control (BACC) Inspection Activities

a. Inspection Scope

The inspectors reviewed the licensees BACC program activities to ensure implementation with commitments made in response to NRC Generic Letter 88-05, Boric

Acid Corrosion of Carbon Steel Reactor Pressure Boundary, and applicable industry guidance documents. Specifically, the inspectors performed an on site record review of procedures and the results of the licensees Mode 3 containment walkdown inspections performed during the Unit 1 Spring 2008 outage. The inspectors also conducted an independent walkdown of the reactor building to evaluate compliance with licensees BACC program requirements and verify that degraded or non-conforming conditions, such as boric acid leaks identified during the Mode 3 containment walkdown, were properly identified and corrected in accordance with the licensees BACC and Corrective Action programs.

The inspectors reviewed the licensees response to a white film found around the annulus of three pressurizer heater sleeves. The licensee conducted radiological and materials characterization testing of the white film and concluded the film was not reactor coolant, and, most likely, from a fiber insulation blanket near the pressurizer heater sleeves.

The inspectors reviewed a sample of engineering evaluations completed for evidence of boric acid found on systems containing borated water to verify that the minimum design code required section thickness had been maintained for the affected components. The inspectors selected the corrosion assessments listed in the Attachment for review:

b. Findings

Introduction:

The NRC identified a finding of very low safety significance (Green)involving a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings for inadequate procedures or compensatory measures associated with conditions that masked indications of boric acid leaks inside containment.

Description:

The inspectors performed a walkdown of containment to independently review the adequacy of the licensees Boric Acid Corrosion Control Program. The inspectors noted a white, crystalline film covering significant portions of containment Levels A and B. While the film was concentrated in the vicinity of, and on, the four emergency core cooling accumulators, it was also found on cable trays, the containment steel liner, decking, various walls, components including valves, insulated piping, bare piping, piping supports, piping flanges and bolting, and electrical junction boxes.

Although the licensee had not determined the composition of this white film, it was concluded to have originated from a mixture of dripping condensation from uninsulated NSCW piping, containment cooler coil leakage of NSCW containing a white calgon additive, dust, and calcium leached from concrete structures inside containment. While no document existed to confirm the original date these conditions began to exist, interviews with plant personnel suggest they have existed since commercial operation (1987).

The licensee took eight random representative samples of the white residue for analysis to determine if the residue potentially masked boric acid leakage. As a result of the sampling, radioisotopes consistent with reactor coolant leakage were discovered at three locations. The licensee identified two leaking valves (1-PSV-8708A and 1-1208-X4-901) as possible sources of the radioisotopes. These two valves were subsequently determined to be the source of reactor coolant leakage.

Licensee procedure 00435-C, Boric Acid Corrosion Control Program, Revision 5.2 required walkdowns inside containment to help identify boric acid leakage. These walkdowns included a containment general leak inspection each outage, VT-2 inspections as per the ASME Boiler and Pressure Vessel Code, and additional visual inspections of specifically targeted systems and components. The widespread presence of a white film, which was very similar in appearance to boric acid residue, has diminished the ability of an inspector performing visual exams (i.e. during walkdowns) to detect boric acid leakage as indicated by the presence of previously undiscovered boric acid at sample location #5. No compensatory measures or provisions in the boric acid program procedures existed to provide for an alternate means of verifying white material on a susceptible component was actually boric acid, for example by chemical or radio-isotopic analysis.

Analysis:

The inspectors determined that the licensees failure to provide adequate procedures for the detection of boric acid was more than minor because it affected the procedure quality attribute of the Barrier Integrity Cornerstone in that there were no additional measures taken to discriminate between boric acid leakage and the white residue present in containment which masked actual boric acid leakage. In addition, although indications of actual boron leakage were masked, the operability of the components was not affected. This finding was determined to be of very low safety significance based on the condition would not have resulted in exceeding the Technical Specification (TS) RCS leakage limit or a total loss of mitigation systems safety function.

Enforcement:

10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Contrary to the above, measures were not implemented to compensate for the reduced effectiveness of visual boric acid inspections. Because this violation is of very low safety significance and has been entered into the licensee's corrective actions program, CR 2008104461, this violation is being identified as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy and is identified as NCV 05000424/2008003-01, Inadequate Procedures to Detect Boric Acid Leakage.

.4 Steam Generator (SG) Tube Inspection Activities

a. Inspection Scope

The inspectors reviewed the Unit 1 SGs tube eddy current testing (ECT) examination activities to ensure compliance with TSs, applicable industry standards, SG Program Procedures, and ASME Code Section XI requirements. The inspectors reviewed examination status reports and discussed them with the site lead Level III analyst to ensure that all tubes with relevant indications were appropriately screened for in-situ pressure testing. The inspectors also reviewed the last Condition Monitoring and Operational Assessment report in conjunction with the inspection status reports to assess the licensee prediction capability for maximum tube degradation. In addition, the inspectors reviewed the latest Degradation Assessment report to identify the scope of the inspection and verify it addressed potential degradation areas, plant specific degradation history, and applicable operating experience. The inspectors verified that appropriate inspection scope expansion criteria were applied based on inspection results of active

and new degradation mechanisms. Furthermore, the inspectors reviewed licensee procedures for tube repair by plugging to verify that repair methods were approved and in accordance with Quality Assurance requirements. In relation to the tube repair methods, the inspectors reviewed the licensees implementation of the tube repair criteria to ensure it was consistent with plant TS. The inspectors also reviewed licensee actions in response to primary to secondary leakage; however no primary to secondary leakage was identified during the previous operating cycle. Additionally, the inspectors reviewed documentation to ensure that data analysts, ECT probes, and equipment configurations were qualified to detect the expected types of SG tube degradation.

The inspectors selected a sample of site-specific Examination Technique Specification Sheets (ETSS) to ensure that their qualification was consistent with industry standards.

The inspectors also directly observed data acquisition for tubes R40C62, R39C62, R35C62, R14C106, R15C105, R14C108, R15C107, R15C108, and R16C107 in SG 2; and tubes R48C32, R47C31, R51C32, and R52C32 in SG 3. The inspectors reviewed ECT data with a qualified analyst for tubes R57C44 (SG 1), R1C61 (SG 3), R11C62 (SG 4), and R12C98 (SG 4). Finally, the inspectors discussed with plant personnel their plans to perform foreign object search and recovery (FOSAR) in response to ECT indications of potential loose parts in the secondary side.

b. Findings

Introduction:

An unresolved item (URI) was identified because additional information from the licensee is required to evaluate a potential finding regarding SG tube damage caused by tube pulling activities in SG 4.

Description:

During Unit 1 spring 2008 outage, the licensee planned to remove two straight tube sections in SG 4, locations R11C62 and R12C98. The purpose of this activity was to study certain Eddy Current indications that were obtained in these locations and compare them with the actual tube material condition. When the removal of the tube section in location R11C62 was in process, the activity was interrupted when the pulling equipment slipped and the tube did not move further, indicating an unusual problem for this type of evolution. The vendor performing this activity conducted further investigation and determined that the tube had not been completely cut despite the procedural checks indicating a complete cut. As a result of this activity, several tubes adjacent to tube R11C62 suffered damage in the U-bend and 7th support plate areas.

The damaged tubes were re-inspected with Eddy Current Testing, stabilized, and repaired by plugging. The licensee performed an Operational Assessment to evaluate the post repair condition of the SGs and determined that the repaired tubes, as well as the rest of the tubes left in service, will maintain their structural integrity until the next inspection opportunity. This issue is unresolved pending completion of NRC review and analysis of the final root cause evaluation and is identified as URI 05000424/2008003-02, Steam Generator Tube Damage as a Result of Tube Pulling Activities.

.5 Identification and Resolution of Problems

a. Inspection Scope

The inspectors performed a review of ISI-related problems, including welding, BACC, and SG inspections that were identified by the licensee and entered into the corrective action program as Condition Reports (CRs). The inspectors reviewed the CRs to confirm that the licensee had appropriately described the scope of the problem and had initiated corrective actions. The review also included the licensees consideration and assessment of operating experience events applicable to the plant. The inspectors performed this review to ensure compliance with 10CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requirements. The corrective action documents reviewed by the inspectors are listed in the report attachment.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification

a. Inspection Scope

Resident Quarterly Observation. The inspectors observed operator performance on April 29, during licensed operator simulator training described on simulator exercise guide Dynamic Simulator Scenarios V-RQ-SE-08204. The simulator scenarios covered operator actions resulting from a loss of the normal charging pump, feedwater regulation valve failure, and loss of RCP seal water with an anticipated transient without a trip (ATWT) and main steam line rupture. Documents reviewed are listed in the Attachment.

The inspectors specifically assessed the following areas:

C Correct use of the abnormal and emergency operating procedures C Ability to identify and implement appropriate actions in accordance with the requirements of the technical specifications C Clarity and formality of communications in accordance with procedure 1000-C, Conduct of Operations C Proper control board manipulations including critical operator actions C Quality of supervisory command and control C Effectiveness of the post-evaluation critique

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

The inspectors reviewed one equipment problem to evaluate the effectiveness of the licensees handling of equipment performance problems and to verify the licensees

maintenance efforts met the requirements of 10 CFR 50.65 (the Maintenance Rule) and licensee procedure 50028-C, Engineering Maintenance Rule Implementation. The inspector also reviewed one safety-significant system to verify that the licensees maintenance efforts met the requirements of 10 CFR 50.65 (the Maintenance Rule) and licensee procedure 50028-C, Engineering Maintenance Rule Implementation. The reviews included adequacy of the licensees failure characterization, establishment of performance criteria or 50.65(a)(1) performance goals, and adequacy of corrective actions. Other documents reviewed during this inspection included control room logs, system health reports, the maintenance rule database, and maintenance work orders.

Also, the inspectors interviewed system engineers and the maintenance rule coordinator to assess the accuracy of identified performance deficiencies and extent of condition.

C Unit 1 and Unit 2 steam generator atmospheric relief valves C Unit 1 loop 1 main feedwater regulating valve 1FV0510

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Evaluation

a. Inspection Scope

The inspectors reviewed the following five work activities to verify plant risk was properly assessed by the licensee prior to conducting the activities. The inspectors reviewed risk assessments and risk management controls implemented for these activities to verify they were completed in accordance with procedure 00354-C, Maintenance Scheduling, and 10 CFR 50.65(a)(4). The inspectors also reviewed the CR database to verify that maintenance risk assessment problems were being identified at the appropriate level, entered into the corrective action program, and appropriately resolved.

C Unit 2 operation on the standby auxiliary transformer with the 2A reserve auxiliary transformer out of service.

C Unit 1 operation in midloop conditions C Unit 2 operation during an extended outage of the 2A diesel generator C Unit 1 operation with a vehicle in the low voltage switchyard C Unit 2 extended operation with 2B train containment spray pump out of service

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following five evaluations to verify they met the requirements of procedure NMP-GM-002, Corrective Action Program, and NMP-GM-002-001, Corrective Action Program Instructions. The scope of this inspection included a review

of the technical adequacy of the evaluations, the adequacy of compensatory measures, and the impact on continued plant operation.

C CR 2008103860, Internals of Unit 1 AFW check valve 11302U4061 were found damaged during inspection.

C CR 2008100480, Unit 1 train A NSCW pump number 6 discharge valve did not meet stroke time requirement C CR 2008105651, Linear indication identified on 2A EDG master rod pin bushing C CR 2008106790, Unit 1 Loop 2 main feedwater regulating valve (1FV0520) control system cycling is affecting reactor power.

C CR 2008105335, Unit 2 HVAC supply damper (2PV12670A) damper will not close as required.

b. Findings

No findings of significance were identified.

1R18 Plant Modifications

a. Inspection Scope

Temporary Modifications. The inspectors reviewed the following temporary modifications (TM) and associated 10CFR50.59 screening criteria against the system design bases documentation and procedure 00307-C, Temporary Modifications. The inspectors reviewed implementation, configuration control, post-installation test activities, drawing and procedure updates, and operator awareness for this TM C TM 10623055, Unit 1, Lift incore exit thermocouple leads C TM 10810623, Unit 1, Installation of a modified plug in the CVCS excess letdown heat exchanger pressure control valve (1HV0123).

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing

a. Inspection Scope

The inspectors either observed post-maintenance testing or reviewed the test results for the following five maintenance activities to verify that the testing met the requirements of procedure 29401-C, Work Order Functional Tests, for ensuring equipment operability and functional capability was restored. The inspectors also reviewed the test procedures to verify the acceptance criteria were sufficient to meet the (TS) operability requirements.

C T-ENG-2008-04, Safety Injection Pump B (1-1204-P6-004) functional test C WO-10806129, Train A motor-driven AFW Pump mini-flow valve (FV-5155) breaker functional test C T-ENG-2007-05, Leading Edge Flow Monitor (LEFM) functional test

C WO-2070314201, Train C Battery Room Exhaust Fan B7003 Damper functional test C WO-2081070901, MFW SG #2 MFIV thermal relief valve manifold functional test

b. Findings

No findings of significance were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

The inspectors performed the inspection activities described below for the Unit 1 refueling outage that began on March 16. The inspectors confirmed that, when the licensee removed equipment from service, the licensee maintained defense-in-depth commensurate with the outage risk control plan for key safety functions and applicable technical specifications and that configuration changes due to emergent work and unexpected conditions were controlled in accordance with the outage risk control plan.

Documents reviewed are listed in the Attachment.

C Reviewed reactor coolant system pressure, level, and temperature instruments to verify that the instruments provided accurate indication and that allowances were made for instrumentation errors.

C Reviewed the status and configuration of electrical systems to verify that those systems met technical specification requirements and the licensees outage risk control plan.

C Observed decay heat removal parameters to verify that the system was properly functioning and providing cooling to the core.

C Reviewed system alignments to verify that the flow paths, configurations and alternative means for inventory addition were consistent with the outage risk plan.

C Observed the licensees control of containment penetrations to verify that the requirements of the technical specifications were met.

C Refueling activities for compliance with TS, to verify proper tracking of fuel assemblies from the spent fuel pool to the core, and to verify foreign material exclusion was maintained.

C Containment closure activities, including a detailed containment walkdown prior to startup, to verify no evidence of leakage and that debris had not been left which could affect the performance of the containment sump.

C Heatup and startup activities to verify that TS, license conditions, and other requirements, commitments, and administrative procedure prerequisites for mode changes were met prior to changing modes or plant conditions. Reactor Coolant System (RCS) integrity was verified by reviewing RCS leakage calculations and containment integrity was verified by reviewing the status of containment penetrations and containment isolation valves.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the following five surveillance test procedures and either observed the testing or reviewed test results to verify that testing was conducted in accordance with the procedures and that the acceptance criteria adequately demonstrated that the equipment was operable. Additionally, the inspectors reviewed the CR database to verify that the licensee had adequately identified and implemented appropriate corrective actions for surveillance test problems.

Surveillance Tests C ABW 2080, Black Start Diesel Operating Procedure C ABW 2010, Place Combustion Turbine Unit Online C 14666-1, Train A Diesel Generator and ESFAS Test C 14363-1, Containment Penetration No. 63 PRT Makeup Water Supply Local Leak Rate Test In-Service Tests (IST)

C 14803B-1, Train B CCW Pump and Check Valve IST and Response Time Test

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Controls to Radiologically Significant Areas

a. Inspection Scope

As a result of recent industry operating experience, the inspectors discussed control of keys to Locked High Radiation Areas (LHRA)s and Very High Radiation Areas (VHRA)s, with the Radiation Protection Manager. The inspectors also reviewed recent corrective action documents in the area of LHRA and VHRA key control.

b. Findings

No findings of significance were identified.

4OA1 Performance Indicator (PI) Verification

a. Inspection Scope

The inspectors sampled licensee submittals for the listed PIs during the period from April 1, 2007 through March 31, 2008, for Unit 1 and Unit 2. The inspectors verified the

licensees basis in reporting each data element using the PI definitions and guidance contained in procedures 00163-C, NRC Performance Indicator and Monthly Operating Report Preparation and Submittal.

Barrier Integrity Cornerstone C Reactor Coolant System (RCS) Specific Activity C RCS Leakage The inspectors reviewed Unit 1 and Unit 2 chemistry and operator log entries, the monthly operating reports and monthly PI summary reports to verify that the licensee had accurately submitted the PI data.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Daily Condition Report Review. As required by Inspection Procedure 71152,

Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the licensees corrective action program. This review was accomplished by either attending daily screening meetings that briefly discussed major CRs, or accessing the licensees computerized corrective action database and reviewing each CR that was initiated.

.2 Focused Review

a. Inspection Scope

The inspectors performed a detailed review of the following CRs to verify the full extent of the issue was identified, an appropriate evaluation was performed, and appropriate corrective actions were specified and prioritized. The inspectors evaluated the CR against the licensees corrective action program as delineated in licensee procedure NMP-GM-002, Corrective Action Program, and 10 CFR 50, Appendix B. Documents reviewed are listed in the Attachment.

C CR 2008101980, Control of the differential pressure across the ECCS sump recirculation valves

b. Findings and Observations

No findings of significance were identified.

.3 Semi-Annual Trend Review

The inspectors performed a review of the licensees Corrective Action Program and associated documents to identify trends which could indicate the existence of a more significant safety issue. The review was focused on repetitive equipment issues, but also

considered the results of inspector daily CR screening and the licensees trending efforts.

The review nominally considered the six month period of January 2008 through June 2008 although some examples extended beyond those dates when the scope of the trend warranted. The inspectors also reviewed several CRs associated with operability determinations which occurred during the period. The inspectors compared and contrasted their results with the results contained in the licensees two latest Integrated Performance Assessments (IPAs). Corrective actions associated with a sample of the issues identified in the licensees trend reports were reviewed for adequacy. The inspectors also evaluated the trend reports against the requirements of the licensees corrective action program as specified in licensee procedure NMP-GM-002, Corrective Action Program, and 10 CFR 50, Appendix B. Documents reviewed are listed in the

.

b. Findings and Observations

No findings of significance were identified. The inspectors compared the licensee IPA with the results of the inspectors daily screening and did not identify any discrepancies or potential trends in the data that the licensee had failed to identify.

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with licensee security procedures and regulatory requirements relating to nuclear plant security. These observations took place during both normal and off-normal plant working hours.

b. Findings and Observations

No findings of significance were identified.

.2 (Closed) Unresolved Item (URI)05000424/2006005-01, Measured Axial Offset is higher

than assumed in the COLR During the review of condition report 2006113450 the resident inspectors noted that measured steady state core axial offset (AO) was outside the analyzed AO band used in the development of the Heat Flux Hot Channel Factor (FQ(Z)) portion of the Plant Vogtles Unit 1 Cycle 14 Core Operating Limits Report (COLR). To justify continued operation, the licensee used the methodology described in Westinghouse report PCT-05-529, Revision 3, Axial Offset Validity Criteria for Westinghouse Reload Safety Analysis Methodology, Nuclear Design Plant Operating Data Predictions, and Power Distribution Measurements including W(z) Factors and Fxy(z) Limits (AO Validity Criteria)to apply a penalty to the Transient FQ(Z) margin. The W(z) values used to determine the transient FQ(Z) were based on an interpolation between pre-selected curves for discrete

core burnup values rather than using a single curve. The resident inspectors raised questions about the use of a limited analysis band for AO, the AO Validity Criteria, and interpolation between burnup-dependent W(z) curves.

After a review of the technical information, the NRC staff determined the use of a limited analysis band for AO was consistent with Westinghouse technical guidance previously approved by the NRC. Additionally, although Westinghouse report PCT-05-529 had not been formally submitted to the NRC for technical review, the NRC staff did not identify a significant safety concern with the use of this report. The staff also determined there was reasonable assurance that the use of different W(z) curves depending on core burnup was consistent with the Vogtle Unit 1 Technical Specifications despite the lack of specificity in the technical report regarding the development, control, and use of burnup dependent W(z) curves. The interpolation between two W(z) curves based on different burnups was considered consistent with the Vogtle Unit 1 Technical Specifications. No violation of NRC requirements was identified.

.3 (Closed) Temporary Instruction (TI) 2515/166, Pressurized Water Reactor Containment

Sump Blockage (NRC Generic Letter 2004-02) Units 1 and 2

a. Inspection Scope

The inspectors performed an in-office review to assess the status of GL 2004-02 commitment actions which were not complete during the previous Units 1 and 2 TI 2515/166 inspections (NRC Report Nos. 50-424,425/2006005, 50-424,425/2007002 and 50-424,425/2007003). These action items included installation of Unit 1 high head safety injection (HHSI) line modifications and flow balance testing of the Unit 2 HHSI lines which were modified in the previous Unit 2 refueling outage.

The inspector reviewed the design change and work documentation related to the Unit 2 HHSI flow balance testing and Unit 1 HHSI modification to verify the activities were complete and accomplished consistent with regulatory design control and 10 CFR 50.59 requirements. The licensee had an approved extension for completion of the Unit 1 HHSI modification until the 2008 spring outage (Letter, USNRC to VEGP, dated December 19, 2007).

b. Findings

No findings of significance were identified. The Units 1 and 2 modifications and program changes related to GL 2004-02 actions were complete and implemented in accordance with design control and 10 CFR 50.59 regulatory requirements.

The incomplete GL 2004-02 commitment for Units 1 and 2 was related to design and licensing documentation. The chemical effects and downstream effects analysis was in progress and had received an NRC approval (NRC letter dated May 29, 2008) for completion date extension to July 31, 2008. An additional item in progress, related to GL 2004-02 corrective actions, was the revision of the licensees Technical Specifications values for minimum refueling water storage tank (RWST) volume and semi-automatic RWST-to- containment sump swap over. The May 29, 2008, letter approved the completion schedule extension for this item, related to procedure changes and training,

to 30 days after the license amendment request (LAR) approval. The LAR was submitted to the NRC on January 9, 2008.

This documentation of TI-2515/166 completion, as well as any results of sampling audits of licensee actions, will be reviewed by the NRC staff (Office of Nuclear Reactor Regulation - NRR) as input along with the Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation During Design Basis Accidents at Pressurized-Water Reactors, responses to support closure of GL 2004-02 and Generic Safety Issue (GSI)-191, Assessment of Debris Accumulation on Pressurized-Water Reactor (PWR) Sump Performance." The NRC will notify each licensee by letter of the results of the overall assessment as to whether GSI-191 and GL 2004-02 have been satisfactorily addressed at that licensees plant(s). Completion of TI-2515/166 does not necessarily indicate that a licensee has finished all testing and analyses needed to demonstrate the adequacy of their modifications and procedure changes. Licensees may also have obtained approval of plant-specific extensions that allow for later implementation of plant modifications. Licensees planned to confirm completion of all corrective actions to the NRC. The NRC will track all such yet-to-be-performed items related to TI-2515/166 which may include future inspections.

.4 NRC Temporary Instruction (TI) 2515/172, Reactor Coolant System Dissimilar Metal Butt

Welds (DMBWs)

a. Inspection Scope

From March 31 to April 4, 2008 and April 22 to 24, 2008, the inspectors reviewed the licensees activities related to the inspection and mitigation of dissimilar metal butt welds in the Reactor Coolant System (RCS) to ensure that the licensee activities were consistent with the industry requirements established in the Materials and Reliability Program (MRP) document MRP-139, Primary System Piping Butt Weld Inspection and Evaluation Guidelines, July 2005. The inspectors activities took place during a refueling outage, and during a second week after the outage, and covered the following:

a) implementation of actions required as part of Confirmatory Action Letter (CAL) No.

NRR-07-005, b) documentation and direct observation of the weld overlay process on the pressurizer surge line nozzle N6, c) documentation review and direct observation of the volumetric examination on pressurizer safety nozzle N2 after completion of the full structural weld overlay (FSWOL), d) review of the MRP-139 program.

b. Findings and Observations

No findings of significance were identified.

MRP-139 Baseline Inspections 1) Have the baseline inspections been performed or are they scheduled to be performed in accordance with MRP-139 guidance? Were the baseline inspections of the pressurizer temperature DMBWs of the nine plants listed in TI 2515/172, 03.01.b completed during the spring 2008 outages?

Yes, all baseline volumetric inspection activities required to be completed per MRP-139 Section 1.2 at the time of this report have been completed.

In addition, all baseline exams required to be completed through December 31, 2010, have been completed.

All Alloy 82/182 butt welds associated with the pressurizer and exposed to pressurizer-like temperatures had FSWOLs completed in the spring 2008 outage at Vogtle Unit 1, and in the spring of 2007 at Vogtle Unit 2. The six welds receiving a FSWOL were the surge line, spray line, one relief, and three safety line nozzles.

There are no Alloy 82/182 butt welds greater than or equal to 4 NPS and less than 14 NPS exposed to temperatures equivalent to the hot leg.

Alloy 82/182 butt welds greater than 14 NPS exposed to temperatures equivalent to the hot leg were inspected during the most recent 10-year American Society of Mechanical Engineers (ASME) code exams. For Unit 1 this was 1R13 (fall 2006),and for Unit 2 this was 2R12 (spring 2007). The welds within this grouping are all four hot leg reactor vessel nozzles.

Alloy 82/182 butt welds exposed to temperatures equivalent to the cold leg were inspected during the most recent 10-year AMSE Code exams. For Unit 1 this was

1R13 (fall 2006), and for Unit 2 this was 2R12 (spring 2007). The welds within this

grouping are all four cold leg reactor vessel nozzles.

2) Is the licensee planning to take any deviations from MRP-139 requirements?

No, the licensee has not submitted any requests for deviation from MRP-139 requirements.

Volumetric Examinations 1) For each examination inspected, was the activity performed in accordance with the examination guidelines in MRP-139, Section 5.1, for unmitigated welds or mechanical stress improved welds and consistent with NRC staff relief request authorization for overlaid welds?

Unit 1 Pressurizer Safety Nozzle N2 Dissimilar Metal Butt Weld (DMBW) After Mitigation by Full Structural Weld Overlay (FSWOL)

Yes, the volumetric examination of the PRZ Safety Nozzle N2 was performed in accordance with a qualified procedure for UT examination, consistent with MRP-139 requirements and the proposed alternative submitted for NRC approval (Proposed Alternative ISI-GEN-ALT-07-01, Version 2, ADAMS Accession Number ML073610061).

The procedure was qualified in accordance with ASME Section XI, Appendix VIII, as implemented through the Electrical Power Research Institute (EPRI) Performance Demonstration Initiative (PDI) Program. Prior to the examination, the licensee verified the FSWOL surface flatness to ensure it permitted volumetric examination as well as the surface roughness to ensure it was 250 µ-inches RMS or better. The licensee conducted the examination 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the third weld layer was

completed. The licensee utilized conventional manual UT technology to perform the examination using procedure PDI-UT-8, Generic Procedure for the Ultrasonic Examination of the Weld Overlaid Similar and Dissimilar Metal Welds, Revision F.

The UT examiners scanned the FSWOL to the maximum extent practicable in two axial and two circumferential directions. The licensee was able to obtain adequate coverage in the UT examination performed to detect fabrication flaws in the FSWOL.

This part of the examination resulted in 98.6% coverage with the 0° angle transducer, 94.6% coverage in the circumferential beam direction, and 100%

coverage in the axial beam direction. For the pre-service (PSI) examination of the new volume above the dissimilar metal weld, the licensee obtained 100% coverage in the circumferential and axial beam directions. For the PSI examination of the new volume above the stainless steel weld, the licensee obtained 100% coverage in the circumferential beam direction and 99.1% coverage in the axial beam direction.

2) For each examination inspected, was the activity performed by qualified personnel?

Unit 1 Pressurizer Safety Nozzle N2 DMBW After Mitigation by FSWOL Yes, the personnel involved in the UT examinations of the PRZ Safety Nozzle N2 FSWOL were qualified in accordance with MRP-139 requirements and the licensees proposed alternative. The examiners were qualified Level II in the UT method as required by the UT procedure and in accordance with the vendors written practice for NDE personnel. The UT examiners were also PDI qualified for the specific UT procedure they implemented. The final examination report was reviewed by a vendors Level II in the UT method and a licensees Level III in the UT method.

3) For each examination inspected, was the activity performed such that deficiencies were identified, dispositioned, and resolved?

Unit 1 Pressurizer Safety Nozzle N2 DMBW After Mitigation by FSWOL Yes, the inspectors reviewed documentation and directly observed field work to verify that deficiencies were identified, dispositioned, and resolved. Based on the inspection activities, the inspectors determined that the examination was conducted in a manner such that deficiencies were identified, dispositioned, and resolved.

Weld Overlays 1) For each weld overlay inspected, was the activity performed in accordance with ASME Code welding requirements and consistent with NRC staff relief requests authorizations? Has the licensee submitted a relief request and obtained NRR staff authorizations to install weld overlays?

Unit 1 Pressurizer Surge Nozzle N6 DMBW FSWOL Yes, the licensee installed Unit 1 pressurizer surge nozzle N6 DMBW FSWOL in accordance with the applicable sections of the ASME Boiler and Pressure Vessel Code (ASME Code). The licensee sought approval for a proposed alternative to certain ASME Code requirements through ISI-GEN-ALT-07-01, Version 2.0.

Approval for the proposed alternative was obtained and an NRC safety evaluation report (SER) was issued, ADAMS Accession Number ML080580291.

The inspectors reviewed welding procedure specifications, procedure qualification records, weld wire certifications, and the in-process welding process control sheets for compliance to ASME Section IX requirements and adherence to the SER. The inspectors also evaluated a number of the licensees corrective action program documents (condition reports), and third party contractor corrective action process issue reports regarding weld overlay quality issues.

2) For each weld overlay inspected, was the activity performed by qualified personnel?

Pressurizer Surge Nozzle N6 DMBW FSWOL Yes, welding personnel were qualified in accordance with the requirements identified in ASME Code Section IX. The inspectors reviewed the welder performance qualification test records and compared them with the requirements of QW-300. The in-process welding process control sheets were reviewed for compliance with the proposed alternative and ASME Code Section IX requirements.

3) For each weld overlay inspected, was the activity performed such that deficiencies were identified, dispositioned, and resolved?

Pressurizer Surge Nozzle N6 DMBW FSWOL Yes, the inspectors reviewed documentation and directly observed field work to verify that deficiencies were identified, dispositioned, and resolved. Based on inspection activities, the inspectors determined that the installation of the FSWOL was conducted in a manner such that deficiencies were identified, dispositioned, and resolved.

Mechanical Stress Improvement (Not Applicable)

The licensee has not implemented Mechanical Stress Improvement as a mitigation method for DMBWs.

In-service Inspection Program 1) Has licensee prepared an MRP-139 in-service inspection program?

No, the licensee did not have a stand alone MRP-139 in-service inspection program document. The licensees MRP-139 inspection program consisted of the documents listed below, which were previously prepared documents, and the inclusion of MRP-139 requirements as augmented inspections in the ASME Section XI In-service Inspection Program (ISI Program). The inspectors reviewed the following documents and held discussions with licensee representatives.

• NMP-ES-029, SNC Alloy 600 Program, Version 2.0

• NMP-ES-029-GL01, Alloy 600 Program Strategic Plan, Version 2.

• Procedure Number 84008-C, RCS Alloy 600 Material Inspection Program, Rev 4.1

• Vogtle Unit 1 Third Interval ISI Plan

• Vogtle Unit 2 Third Interval ISI Plan 2) Are welds appropriately categorized?

The inspectors reviewed all welds categorized at the time of the inspection for appropriate categorization in accordance with MRP-139, Section 6. With one exception, welds were appropriately categorized.

The pressurizer nozzles were correctly categorized as Category D welds in their pre-FSWOL condition. However, the Third Interval ISI Plan incorrectly stated that those welds would be removed from the scope of the MPR-139 program following the application of the FSWOL. Those welds should be re-categorized as Category B or F following the application of the FSWOL. This was brought to the attention of the licensee who entered the issue into their Corrective Action Program as CR 200810526. Because all overlaid pressurizer nozzle welds were scheduled to be examined in the outage following their application (approximately 18 months from the date of this report), as a commitment made through proposed alternative ISI-GEN-ALT-07-01, Version 2.0 and the associated SER, the licensee would have met the most restrictive examination extent and schedules (Category F) which required an examination once in the next five years.

3) Are inspection frequencies consistent with the requirements of MRP-139?

Yes, planned inspection frequencies for welds in the MRP-139 program are consistent with the requirements of MRP-139.

4) What is the licensees basis for categorizing welds as H or I and plans for addressing potential PWSCC?

No welds were categorized as Categories H or I.

5) What deviations has the licensee incorporated and what approval process was used?

No deviations to MRP-139 have been incorporated by the licensee.

4OA6 Meetings, Including Exit

.1 Exit Meeting

On July 21, the resident inspectors presented the inspection results to Mr. C. R.

Dedrickson and other members of his staff, who acknowledged the findings. The inspectors confirmed that proprietary information was not provided or examined during the inspection.

4OA7 Licensee Identified Violations

The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited Violation.

C TS 5.7.2 requires that areas with radiation levels greater than or equal to 1000 mrem/hr be locked and the keys be maintained under the administrative control of Operations or health physics (HP) supervision. Contrary to this, on March 31, 2008, it was discovered that Security had master keys to Locked High Radiation Areas (LHRAs) that were not under the administrative control of Operations or HP supervision. This violation was discovered by HP supervisors performing an audit of key controls based on recent industry operating experience. Immediate corrective actions were taken upon discovery and documented in CR 2008103892. Although this event involved failure to maintain proper controls to LHRAs, this finding is of very low safety significance because there was no evidence of unauthorized worker entry into the affected areas nor any unexpected radiation exposures to licensee personnel.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

R. Brown, Training and Emergency Preparedness Manager

C. Buck, Chemistry Manager

W. Copeland, Performance Analysis Supervisor

R. Dedrickson, Plant Manager

K. Dyar, Security Manager

I. Kochery, Health Physics Manager

J. Robinson, Work Control Superintendent

T. Tynan, Site Vice-President

D. Vineyard, Operations Manager

J. Williams, Site Support Manager

T. Youngblood, Site Engineering Manager

NRC personnel

S. Shaeffer, Chief, Region II Reactor Projects Branch 2

LIST OF ITEMS

OPENED AND CLOSED

Opened

05000424/2008003-02 URI Steam Generator Tube Damage as a Result of Tube Pulling Activities (Section 1R08)

Opened and Closed

05000424/2008003-01 NCV Inadequate Procedures Associated with Degraded Ability to Detect Boric Acid Leakage (Section 1R08)

Closed

05000424/2006005-01 URI Measured Axial Offset is higher than assumed in the COLR (Section 4OA5.2)

2515/166 TI Pressurized Water Reactor Containment Sump Blockage (NRC Generic Letter 2004-02) Units 1 and 2 (Section 4OA5.3)

Discussed

2515/172 TI Reactor Coolant System Dissimilar Metal Butt Welds (DMBWs) (Section 4OA5.4)